In the field of nuclear power facilities, it is important to analyze the reactor core neutronic properties for maintaining the nuclear power facility and training reactor operators to perform routine and emergency monitoring procedures. Heretofore, core neutronics for a light water reactor have been analyzed using engineering codes, such as the coarse mesh method described in Borresen, "A simplified, Coarse-Mesh, Three-Dimensional Diffusion Scheme for Calculating the Gross Power Distribution in a Boiling Water Reactor," Nucl. Sci. Engr., 44, 37, 1971, and methods of the RAMONA-3B code described in Wulff et al., "A Description and Assessment of RAMONA-3B Mod. O Cycle 4: A Computer Code with Three-Dimensional Neutron Kinetics for BWR System Transients," NUREG/CR-3664, Brookhaven National Laboratory, January 1984. These codes provide methods for providing a set of core neutronics parameters in a defined circumstance and analyzing or determining the resultant reactor core neutronics parameters in response to the given conditions.
The Borresen reference refers to obtaining core neutronics data and solving modified two-group neutron diffusion equations for the two types of neutrons inside the core, namely the fast neutrons and the thermal neutrons. A thermal neutron may be considered as a fast neutron that has slowed down. More specifically, the reactor core is represented as a number of nodes that are spaced apart such that the fast neutrons have a relatively large mean free path (i.e., diffusion length) and the thermal neutrons have a low leakage from node to node. This permits using an approximation for the thermal neutrons leakage and a modification of the two-group equations to simplify the number of steps required to determine the core neutronics for the given conditions.
RAMONA-3B, developed by the Brookhaven National Laboratory, uses the Borresen coarse mesh method and also relies on the fast neutrons as the determining criteria. However, the RAMONA-3B method relies on solving the two-group neutron diffusion equations by a finite difference method to determine the core neutronics for the given conditions.
One of the problems with these known techniques is that the model does not have the capability to run the code from power plant start up to shutdown continuously in real time. They do not have the ability to analyze dynamic or static conditions in real-time. Consequently, they are limited in their application to selected transient conditions. Further, those known techniques are not sufficiently flexible to train operators under a wide variety of conditions or in real-time environments.
It is therefore, an object of the present invention to provide for determining core neutronics in a real-time environment. It is another object to provide for a real-time analysis of core neutronics that can be used for simulation training of facility operators and for engineering analysis of core neutronics, separately or simultaneously.
It is another object of the invention to provide for determining core neutronics in response to rapid transient conditions in a real-time environment.
It is another object of the invention to provide a real-time analysis of core neutronics under normal and emergency operating conditions.
It is another object of the invention to simulate real-time core neutronics under normal, emergency, and beyond design conditions continuously.